Model development for analysis of the Korea advanced liquid metal reactor

Chang, Won-Pyo; Kwon, Young-Hyuk; Lee, Yongbum; Hahn, Dohee

doi:10.1016/s0029-5493(02)00129-2

Cited by 20 publications

(2 citation statements)

References 5 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Studies have been reported in connection with the natural convection decay heat removal of the KALIMER reactor under design in Korea (9). As a first part of the numerical study, a simple closed sodium circulation loop is simulated for the evaluation of the formation and effectiveness of natural circulation in a liquid metal loop.…”

Section: Analysis Of a Natural Circulation Loopmentioning

confidence: 99%

Natural Convection in Secondary Sodium Circuit of Fast Breeder Test Reactor

Vaidyanathan¹,

Kasinathan²,

Velusamy³

2011

Int. Jour. Des. Manf. Tech.

View full text Add to dashboard Cite

Fast Breeder Test Reactor(FBTR) is in operation since 1985. Tests and analysis have been performed to assess the conduciveness of the layout to natural convection under loss of pumping flow conditions. One such test refers to the secondary sodium circuit, which forms the link between the heat source (Core) in the primary sodium system and the steam generator which is the final heat sink. This paper details out the tests and two one dimensional numerical models for assessing the flow and compares the predictions with actual test results. Though there is a reasonable match, better assessment is possible with multidimensional models to predict the flow stagnation regions.

show abstract

Section: Analysis Of a Natural Circulation Loopmentioning

confidence: 99%

Natural Convection in Secondary Sodium Circuit of Fast Breeder Test Reactor

Vaidyanathan¹,

Kasinathan²,

Velusamy³

2011

Int. Jour. Des. Manf. Tech.

View full text Add to dashboard Cite

show abstract

“…JOYO and MONJU in Japan referred in [1] belong to loop-type designed fast reactor. Compared to the looptype design, the pool-type design described in [2] Many analysis codes have been developed for evaluating the thermal-hydraulic performance in DRACS, such as SASSYS described in [3] and COMMIX described in [4] Schematic diagram of DRACS developed in USA, SSC-K in Korea(described in [5]), further development based on SSC-L in Japan( described in [1] The inlet and outlet in the primary circuit of DRACS in pool-type FR are immersed in hot sodium pool; the intermediate circuit is composed of the tube side of DHX and AHX, together with the corresponding pipes. The momentum equation in the closed circuits can be expressed as follows:…”

Section: Introductionmentioning

confidence: 99%

Development of passive direct reactor auxiliary cooling system model in Liquid Metal Fast Breeder Reactor

Sui

Guo

et al. 2013

2013 International Conference on Materials for Renewable Energy and Environment

View full text Add to dashboard Cite

Passive safety is a design concept for fast reactors. An analysis code to perform thermal-hydraulic calculation in passive Direct Reactor Auxiliary Cooling System (DRACS) was developed through the coupling of thermalhydraulic response in Direct Heat EXchanger (DHX) model, Air Heat EXchanger (AHX) and circuits of DRACS. After steady-state initialization, a transient starting from air damper opening at the instant of reactor scram was simulated. Key parameters, such as mass flow rate in intermediate loop under natural circulation, heat removed by DRACS, air temperature at AHX outlet, coolant temperature at both side of DHX, are calculated, which preliminary showed the effectiveness of the code. As part of the system analysis code development for the pool-type Liquid Metal Fast Breeder Reactor (LMFBR), the newly developed code is now ready to be incorporated into System Analysis Code for pool-type Fast Reactor (SAC-CFR) to perform the interactive response between the DRACS and the whole system after scram.

show abstract